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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Distribution Development for STORM Ingestion Input Parameters

The Sandia-developed Transport of Radioactive Materials (STORM) code suite is used as part of the Radioisotope Power System Launch Safety (RPSLS) program to perform statistical modeling of the consequences due to release of radioactive material given a launch accident. As part of this modeling, STORM samples input parameters from probability distributions with some parameters treated as constants. This report described the work done to convert four of these constant inputs (Consumption Rate, Average Crop Yield, Cropland to Landuse Database Ratio, and Crop Uptake Factor) to sampled values. Consumption rate changed from a constant value of 557.68 kg / yr to a normal distribution with a mean of 102.96 kg / yr and a standard deviation of 2.65 kg / yr. Meanwhile, Average Crop Yield changed from a constant value of 3.783 kg edible / m 2 to a normal distribution with a mean of 3.23 kg edible / m 2 and a standard deviation of 0.442 kg edible / m 2 . The Cropland to Landuse Database ratio changed from a constant value of 0.0996 (9.96%) to a normal distribution with a mean value of 0.0312 (3.12%) and a standard deviation of 0.00292 (0.29%). Finally the crop uptake factor changed from a constant value of 6.37e -4 (Bq crop /kg)/(Bq soil /kg) to a lognormal distribution with a geometric mean value of 3.38e -4 (Bq crop /kg)/(Bq soil /kg) and a standard deviation value of 3.33 (Bq crop /kg)/(Bq soil /kg)

60 APPLIED LIFE SCIENCES↗
DOE OSTI2013

Joint IAEA/NNSA International Workshop Nuclear Forensics Methodologies for Practitioners 2013 Scenario Based Exercise – Version 4.0 Instructor’s Manual

[Participants will serve as border guards for Reimerland. They will be given brief instruction on the operation of hand-held RadioIsotope DetectorS (RIDS) and be provided an intelligence briefing that tells them to be on the lookout for suspicious activity at their post. Their instruction will include directing suspicious vehicles to a location for secondary screening. If, after secondary screening, suspicions of a criminal act involving nuclear and or radioactive materials remain, participants have been instructed to request assistance from the NLEA, who will then setup and manage a radiological crime scene. Participants will watch a demonstration of two vehicles containing radioactive materials driving through and setting off a portal monitor. The first vehicle, a semi-tractor trailer, sets off only a gamma alarm. After the driver provides a shipping manifest of fertilizer, participants, posing as border guards, are expected to waive this vehicle through inspection. The second vehicle, an SUV, set off both gamma and 2 neutron alarms. The alarming of the neutron monitor should prompt participants to set up a secondary inspection of the vehicle immediately. The driver of the vehicle indicates he is in legal possession of an industrial instrument containing an old 133Ba source that has decayed to a level no longer requiring official paperwork according to the IAEA and internationally accepted transportation regulations. Authorities have verified that the industrial source does fit the description of one that is sold commercially. However, upon setting up a secondary screening, participants will use hand-held detectors to locate several other radioactive sources emanating from a black duffle bag in the rear of the vehicle (Figure 1). Hand held detectors detect the presence of 133Ba, and Pu. Upon questioning, the driver only commits to having the 133Ba industrial source and cannot account for the detection of neutrons within his vehicle. Since neutron alarms also sounded, participants should indicate that a neutron alarm would be inconsistent with a 133Ba source alone and should therefore conclude further investigation is warranted. This will prompt participants to call in a response team from the NLEA to set up a radiological crime scene around the vehicle in question. The response team is able to shoot a 3-D X-ray radiograph of the duffle bag without moving it to ensure it is rendered safe and moveable without disturbing the contents in the field (Figure 2). At this point, the duffle bag is entered into inventory as evidence and a chain of custody form is initiated. Swipes are taken from the outer bag to confirm there is no dispersible contamination. The bag and its contents are considered valuable for the investigation by the lead investigator. He determines the duffle bag is safe to transport to RRL for evidence inventory and analysis. The duffle bag and its contents are packaged and sent off to the RRL.]

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2023

Development of local-power-free, remote α -particle detection using optical fibers

We demonstrate the application of fluorescence optical fiber coupled to a telecom grade fiber as a sensor for alpha particles using alpha-specific ZnS(Ag) scintillation materials whose wavelength is down-shifted into a low-loss region of the telecom grade fiber transmission band. Telecom-grade fiber optics offer a solution for sensing alpha radiation in deep repositories and cask storage for radioactive materials due to the stability of SiO 2 under normal environmental conditions and its relative radiation hardness at low radiation doses. Long-term nuclear waste storage facilities require sensors for the detection of leakage of radioactive materials that are maintenance-free, do not require power and can survive with no ‘wear out’ mechanisms for decades. By accomplishing the wavelength transformation, we maximize efficiencies in the detection of α-particles and signal transport and can detect alpha scintillation at distances on the order of >1 km with a sensor that is ~3% efficient and can be easily scaled as a sensor array. This paper describes the construction and testing of the sensor including manufacture of the controlled thickness films, verification of the wavelength shift from 450 to 620 nm and optimization of the sensitivity as a function of thickness. We also model the relative sensitivity of the film as a function of film thickness, and we demonstrate a signal-to-noise ratio of 10 at a range of greater than 1 km.

alpha particles↗
DOE OSTI2024

Assessment of Potential Dose and Environmental Impacts from Proposed Testing at the INL Radiological Response Training Range

This assessment uses screening level models to calculate potential environmental impacts from proposed tests at the Idaho National Laboratory (INL) Radiological Response Training Range (RRTR) site. Proposed tests could be conducted using 11 different radioactive material types that include K 2 O, LaBr 3 , KBr, Cu, Zr, F, Ga, Ga 2 O 3 , NaNO 2 , Ga-68, and Tc-99m. The tests could potentially release radioactive material to the atmosphere and radionuclides and other contaminants to the soil, which could leach into the unsaturated zone and migrate to the aquifer. Atmospheric transport of radionuclides to potential human receptors and time-integrated air concentrations were calculated with a Gaussian plume model and three years of hourly meteorological data. Potential surface-soil impacts were calculated with the computer program Mixing-Cell Model (MCM). Groundwater impacts were calculated with the computer programs MCM and GWSCREEN. Radiological doses from potential atmospheric releases were calculated for public receptors off the INL Site and for workers at nearby INL facilities. Results were compared to regulatory dose limits. Maximum potential groundwater concentrations were estimated in the aquifer below the NSTR site and compared to drinking water standards or risk-based screening levels for resident tap water. Soil concentrations were calculated and compared to risk-based screening levels for workers and potential future residents. All impacts were estimated based on the assumption that 12 tests are conducted annually using all 11 material types for a period of 15 years. This document provides the resources to enable a subject matter expert in the field of environmental assessments to replicate the modeling and calculations. The methodology and parameters are presented in the text. All electronic files, including computer code input, output, executable files, batch files, scripts, and spreadsheet files are contained in a zip file that can be accessed by selecting “Additional Information” (select Native File) in the INL Electronic Document Management System (EDMS).

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Assessment of Potential Dose and Environmental Impacts from Proposed Testing at the INL National Security Test Range

This assessment uses screening level models to calculate potential environmental impacts from proposed tests at two locations at the Idaho National Laboratory (INL) National Security Test Range (NSTR) site. Proposed tests could be conducted using 11 different radioactive material types that include K 2 O, LaBr 3 , KBr, Cu, Zr, F, Ga, Ga 2 O 3 , NaNO 2 , Ga-68, and Tc-99m. The tests could potentially release radioactive material to the atmosphere and radionuclides and other contaminants to the soil, which could leach into the unsaturated zone and migrate to the aquifer. Atmospheric transport of radionuclides to potential human receptors and time-integrated air concentrations were calculated with a Gaussian plume model and three years of hourly meteorological data. Potential surface soil impacts were calculated with the computer program Mixing-Cell Model (MCM). Groundwater impacts were calculated with the computer programs MCM and GWSCREEN. Radiological doses from potential atmospheric releases were calculated for public receptors off the INL Site and for workers at nearby INL facilities. Results were compared to regulatory dose limits. Maximum potential groundwater concentrations were estimated in the aquifer below the NSTR site and compared to drinking water standards or risk-based screening levels for resident tap water. Soil concentrations were calculated and compared to risk-based screening levels for workers and potential future residents. All impacts were estimated based on the assumption that 12 tests are conducted annually using all 11 material types for 15 years. This document provides the resources to enable a subject matter expert in the field of environmental assessments to replicate the modeling and calculations. The methodology and parameters are presented in the text. All electronic files, including computer code input, output, executable files, batch files, scripts, and spreadsheet files, are contained in a zip file that can be accessed by selecting “Additional Information” (select Native File) in the INL Electronic Document Management System (EDMS).

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2018

The DISTINCTIVE University Consortium: Transforming High Activity Materials Research in the UK - 18212

The DISTINCTIVE consortium (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories) is a partnership of 11 research universities, working to decommission and clean up the legacy of civil nuclear energy generation, sponsored by the Engineering and Physical Sciences Research Council (EPSRC). An important component of the research programme is to develop the skills and facilities to address key research questions, requiring the use of highly radioactive materials. Here we highlight the outcomes of four research projects in which access to state of the art research facilities for working with radioactive materials, supported by international partnerships, has advanced research in plutonium storage and immobilisation, and the disposal of spent nuclear fuels. First, we have developed a viable formulation and process for immobilisation of the UK plutonium stockpile, including commissioning of a radiological hot isostatic press and validation studies using plutonium-239. Second, we have further underpinned the interim storage of PuO{sub 2}, through understanding of water and chlorine sorption / desorption, pertinent to the historic practice of containment within, and degradation of PVC bags. Third, we have developed a novel methodology to investigate the in situ radiolysis induced dissolution of actinide oxide thin films, using synchrotron radiation, to underpin the geological disposal of spent nuclear fuel. Fourth we monitored in situ the oxidation and self-ignition mechanism of uranium carbide, a legacy fuel from the Dounreay Prototype Fast Reactor that is waiting for oxidation treatment prior to permanent storage and/or disposal. Through this programme of research, the DISTINCTIVE consortium is rebuilding the UK capability for research with transuranic materials, which, hitherto, was eroded over decades of disinvestment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Is Entombment an Acceptable Option for Decommissioning? An International Perspective - 13488

Selection of a decommissioning strategy is one of the key steps in the preparation for decommissioning of nuclear facilities and other facilities using radioactive material. Approaches being implemented or considered by Member States include immediate dismantling, deferred dismantling and entombment. Other options or slight modifications of these strategies are also possible. Entombment has been identified in the current International Atomic Energy Agency (IAEA) Safety Standards as one of the three basic decommissioning strategies and has been defined as a decommissioning strategy by which radioactive contaminants are encased in a structurally long lived material until radioactivity decays to a level permitting the unrestricted release of the facility, or release with restrictions imposed by the regulatory body. Although all three strategies have been considered, in principle, applicable to all facilities, their application to some facilities may not be appropriate owing to political concerns, safety or environmental requirements, technical considerations, local conditions or financial considerations. The IAEA is currently revising the decommissioning Safety Standards and one of the issues widely discussed has been the applicability of entombment in the context of decommissioning and its general objective to enable removal of regulatory control from the decommissioned facility. The IAEA recently established a consultancy to collect and discuss experience and lessons learned from entombment projects, to identify regulatory requirements and expectations for applying entombment as a decommissioning option strategy, in compliance with the internationally agreed standards. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

CASTOR THTR/AVR Containment Review

The Lawrence Livermore National Laboratory (LLNL)Radioactive Material Packaging and Transportation Review Team (herein referred to as ‘LLNL Staff’) was tasked with the technical review and confirmatory analysis of the containment analysis and corresponding leakage test implementation procedure for the CASTOR®THTR/AVR radioactive material transportation package with the AVR contents.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

The effect of metals on zeolite crystallization kinetics with relevance to nuclear waste glass corrosion

Abstract Geologic disposal of vitrified radioactive material is planned in several countries, but there are remaining uncertainties related to the long-term stability of glass exposed to groundwater. Specifically, the crystallization of aluminosilicate zeolite minerals can accelerate the rate at which glass corrodes and radioactive material is released into the biosphere. In this study, we identify elemental species that may accelerate or suppress zeolite formation using a protocol to examine their effects on zeolite synthesis over a three-day duration. Our results are consistent with previous works demonstrating glass corrosion acceleration in the presence of calcium. Furthermore, we identify two elements—tin and lithium—as inhibitors of zeolite P2 (gismondine, or GIS type) nucleation and, thus, promising components for promoting the long-term durability of glass waste forms.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Summary of the Graduate Mentoring Assistance Program for PNNL PNNL.12.006 (April 2012 – May 31, 2013)

Summary of Students and Projects: Sarah Bender: The possible improvement of the Multi-Isotope Process (MIP) Monitor by the addition of Compton suppression is investigated for this study. The Compton continuum from the dominant 661.7 keV Cs-137 fission product peak obscures lower energy lines which could be used for spectral and multivariate analysis. A novel Compton suppression system which uses fast, room-temperature materials will be developed using simulation. As opposed to the standard high purity germanium (HPGe) detector suppressed using a thallium-doped sodium iodide (NaI(Tl)) guard detector array, cerium-doped lanthanum bromide (LaBr3) will be used as the primary spectroscopic detector. For Sarah’s Master’s degree, she constructed and validated the unsuppressed detector response of the Pennsylvania State University (PSU)-Compton suppression system. For Sarah’s PhD, she is completing that simulation by including the suppressed operation which will be validated with spent fuel samples. She is also going to use the validation of suppressed operation to predict the detector response of a LaBr3 based Compton suppression system. Richard Clark: Richard Clark completed his study this FY of a new technique known as Intrinsic Dosimetry. Intrinsic Dosimetry is the method of measuring total absorbed dose received by the walls of a container holding radioactive material. By considering the total absorbed dose received by a container in tandem with the physical characteristics of the radioactive material housed within that container, this method has the potential to provide enhanced pathway information regarding the history of the container and its radioactive contents. The latest in a series of experiments designed to validate and demonstrate this newly developed tool are reported. Kenneth Dayman: The MIP Monitor is an online, non-destructive approach to process monitoring in a nuclear fuel reprocessing facility using gamma spectra taken of process streams and multivariate data analysis to detect anomalies in operation. Last year, Mr. Dayman investigated methods to classify spent nuclear fuel according to both reactor type and burn up based on the constituent nuclide activities, which may be used to more accurately calibrate other models in an analysis scheme in order to improve the accuracy and precision of predictions. He is currently writing up his results for publication and has transitioned to another project with Dr. Derrick Haas at PNNL.

Schwantes, Jon M.↗
DOE OSTI2021

2021 ROMANIA TRANSPORT SECURITY TABLE-TOP EXERCISE (TTX) – LESSONS LEARNED AND GOOD PRACTICES

During the period of 15-18 March 2021, the IAEA conducted a national workshop addressing the security of radioactive material in transport. The workshop was jointly hosted by the Romanian regulatory authority and the General Inspectorate of Romanian Gendarmerie. The workshop assisted Romania in improving its capability to evaluate the ability of response forces to counter a malicious act against radioactive material in transport. The event consisted of briefings, exercises, a demonstration of transport security equipment, and a transport security table-top exercise (TTX). The workshop was conducted in a virtual-hybrid format in such a way as to meet the COVID-19 related requirements in compliance with Romania’s guidelines. This paper provides an overview of the workshop’s activities, lessons-learned, and good practices related to the transport security TTX exercises.

Maican, Catalin↗
DOE OSTI2019

Expanded Use of ISOCS Characterization Techniques for WIPP Certification at the Idaho Clean-up Project - 19623

The Advanced Mixed Waste Treatment Project (AMWTP) has been retrieving, treating, characterizing, and shipping legacy Transuranic (TRU) waste for disposal since 2003 as part of the overall Idaho Cleanup Project (ICP) in support of the US Department of Energy. As the ICP has progressed, over 700 containers have been identified that cannot be successfully assayed using the current suite of non-destructive assay (NDA) instruments. To meet this challenge, Fluor Idaho, LLC, and Mirion Technologies (formerly Canberra Industries) have teamed to deploy the In-Situ Object Counting System (ISOCS) to supplement the current NDA capabilities. ISOCS has long been used in various radioactive waste management applications, but never has the ISOCS ability to perform intrinsic efficiency modeling of discrete waste containers been employed to characterize and certify TRU waste for disposal at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico. Since the primary disposal facility for this waste is the WIPP, the characterization processes for determining the quantity of radioactive material entrained in the waste are subject to WIPP Waste Acceptance Criteria (WAC) quality requirements and associated performance criteria. Until now, the Carlsbad Field Office (CBFO) WIPP-certified NDA equipment has utilized pre-defined efficiency calibrations for fixed geometries. However, due to the radioactive material content, activity levels, waste stream matrix, and/or the geometries of various size drums, boxes and other-odd-shaped containers or container integrity issues, it may not be suitable to characterize these containers with any of the existing WIPP-certified NDA equipment located at the AMWTP. Therefore, advanced technologies are needed to complete characterization of these problematic waste containers. The July 5, 2016 release of the WIPP WAC, DOE/WIPP-02-3122, Revision 8.0, Appendix A, 'Radioassay Requirements for Transuranic Waste,' was expanded to include several newer concepts such as American Society for Testing and Materials (ASTM) approved methods for efficiency-based calibrations and the use of mathematical computer modeling techniques for calibration. This change prompted AMWTP to pursue obtaining WIPP certification authority for expanded ISOCS techniques to deal with the remaining challenging waste containers and the development and implementation of Mirion's Figures of Merit (FOM) evaluations. These vital FOM evaluations enhance the ISOCS's modeling capabilities by optimizing the dynamic efficiency calibration for each object and reducing the assay's Total Measurement Uncertainty (TMU) budget. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Using the CAP88 Plume Model & Dose Assessment Code for the FTWC Project

The Environmental Protection & Compliance Division (EPC) uses the Clean Air Act Assessment Package – 1988 (CAP88) plume model and dose assessment code to evaluate measured emissions or projected emissions of airborne radionuclides from Los Alamos National Laboratory (LANL) facilities. This software package will use facility data (e.g., stack height and exit velocity), emissions parameters (chemical form and quantity), and environmental data (wind speed and direction) to predict downwind concentrations of radioactive material and subsequent dose to a member of the public at that location. The United States Environmental Protection Agency (EPA) has approved CAP88 for use in demonstrating compliance with the Radionuclide NESHAP2, that portion of the Clean Air Act addressing emissions of airborne radioactive material from Department of Energy facilities. EPC Division has two procedures which discuss the use of CAP88.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

IRAD-MPE Radiological and Nuclear Security Threats (Day 1) [Slides]

This module is designed to enhance awareness regarding radiological and nuclear security threats and the types of radioactive materials that can be used for criminal purposes or terrorist acts. The goals are to: (1) Review the Threat Assessment Process for MPEs, (2) Recognize Potential Radiological and Nuclear Security Threats, (3) Understand the basics of Radiological Dispersal Devices (RDD) and Radiological Exposure Devices (RED), (4) Learn about Medical, Commercial, and Naturally Occurring Radioactive Materials (NORM) sources, and (5) Understand the term Material Out of Regulatory Control (MORC).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Development of the Mobile Systems for Conditioning of Disused Sealed Radioactive Sources in Serbia - 20105

Sealed radioactive sources (SRS) are being used worldwide in the field of medicine, agriculture, industry and research. They can be found in mobile as well as stationary devices. SRS contains radioactive material that is (a) permanently sealed in a capsule or (b) closely bounded and in a solid form. The capsule or material of an SRS should be strong enough to maintain leak tightness under the conditions of use and purpose for which the source was designed, also in case of accidents. In this case only emitted radiation is utilized. Firstly, the hazard from external radiation has to be considered, but the possibility of contamination due to fracture of the capsule should not be disregarded. The radioactive sources are composed of the radiating isotope contained in the filling medium, the single or double isotope holder that partially or totally surrounds the filling medium, the outer cover that contains the parts mentioned above and the capsule closed airtightly by welding or using some other method. The capsule must be tested for leakage periodically. If the SRS is no longer needed (e.g. replaced by a different technique) or it becomes useless for the intended application (e.g. the activity becomes too weak, the equipment containing the source works poorly or becomes obsolete, the source is damaged or leaking) it is considered disused. Disused sealed radioactive sources (DSRS) are typically conditioned and disposed if a facility is available. If the disposal option is not available, conditioned DSRS should be stored under proper conditions. In some cases, the radionuclide(s) in DSRS can be recovered/recycled or the DSRS can be repurposed for other applications. Conditioning of DSRS ensures containment of the radioactive material, provides confinement for leaking sources, provides sufficient radiation shielding, reduces storage/disposal volume by allowing consolidation of multiple sources into a single storage/disposal container, facilitates transport operations and contributes to safety and security as well. Typically, conditioning technologies are deployed either as permanently installed stationary systems in centralized or mobile on-site waste processing facilities, or in a mobile configuration. Centralized stationary facilities provide a single processing location for multiple users that requires transport of the waste to the facility. On the other hand, mobile systems may be provided for the selection and application of the optimum technology for a specific waste stream (such as DSRS) by bringing the process to the point where the waste is generated. In addition, mobile systems could offer additional flexibility by sharing equipment among multiple waste generating sites for processing campaigns that vary in duration, from very short periods to several years. The term 'mobile processing system' refers to any radioactive waste processing system or component which is designed to be transportable and which is not considered permanently installed. Two mobile system for conditioning of disused sealed radioactive sources are developed in the Public Company Nuclear Facilities of Serbia. Development of these mobile systems was supported by SRB9005 national project via Technical Cooperation of the IAEA. The first mobile system, built inside the 20 feet ISO container, will be used for conditioning of DSRS category 3 to 5. The second mobile system, built inside the 7 m long vehicle (Iveco Daily Van), will be used for dismantling of ionizing smoke detectors mostly with Am-241 sources. Designs of the mobile systems were defined in cooperation with two companies from Belgium (Belgoprocess and Leniko) and a Croatian company Ekoteh as well as with the support of the IAEA experts. The generic safety assessment and operational procedures for the mobile systems are developed. Based on safety assessment the acceptance criteria and operational limits and conditions are established. Operational procedures include: (a) equipment and material requirements, (b) assembling procedure of the mobile unit, (c) procedure for acceptance of devices for dismantling and conditioning, (d) dismantling procedure for devices to recover the DSRS, (e) characterization of DSRS, (f) encapsulation procedure of DSRS, (g) disassembling procedure of the mobile unit, and (h) keeping records, identification and traceability. In addition, radiation safety, health safety, security and emergency preparedness plans are prepared. The generic safety assessment and operational procedures could be updated with site specific requirements, DSRS inventory, and different needs for future customers. Developed mobile systems could be used in all situations when it is feasible to perform conditioning of DSRS on the spot in the county and worldwide. Development of these mobile units was just the first step to create the Reference Center for Radioactive Waste Treatment and Disused Radioactive Sources Conditioning for Small Facilities which can become a regional training center in the future, and/or as a tool for comprehensive national search and secure programmes. In the next phases development of e-learning platforms and blended learning packages as well as application for the IAEA Qualified Technical Centre (QTC) for the management of DSRS is foreseen. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Minimum Detectable Quantity Calculation for Radiation Portal Monitors

The minimum detectable quantity of a radiation portal monitor is the smallest amount of radioactive material that can be detected passing through the monitor (with a specified detection probability). The minimum detectable quantity is a function of many factors, including isotope, velocity, distance between pillars, height of portal and vehicle, source distribution and position, background radiation, background suppression, detector volume and positions, detector efficiency, decision metrics and algorithms, the presence of NORM (naturally occurring radioactive material), and alarm thresholds. Experimentally testing a radiation portal monitor to failure for all combinations of these factors is extremely time-consuming and often cost prohibitive. This document outlines a straightforward method for quickly estimating the minimum detectable quantity for moving sources over a large variety of conditions with a minimal number of static measurements. The document also describes a proof of concept software application that the International Atomic Energy Agency has designed based on the described method that can be used by Member States to estimate a monitor’s minimum detectable quantity.

Blessinger, Christopher S.↗
DOE OSTI2022

Cryogenic tracer irradiation facility at the university of Texas at Austin

We report on a cryogenic gas irradiation facility in the 1.1 MW TRIGA reactor at The University of Texas at Austin. The system was designed to produce radioactive xenon and argon for environmental studies, and it can be applied to produce other gaseous radiotracers. The system design includes modeling accident scenarios to ensure operation does not risk damage to the reactor and minimizes risk of release of radioactive material. In conclusion, the SCALE code was used to model a 30-day irradiation of one liter of Xe-126 and the predicted activity is 0.9 Ci of Xe-127 at irradiation end.

Ar-37↗
DOE OSTI2022

General Employee Radiological Training: 24 Month Refresher (Course #47968)

GERT is required by the 10 CFR 835, the Federal Regulation on Radiation Protection that we follow here at LANL. It’s the training for non Rad Workers who may enter Radiological Controlled Areas (RCAs) and Radioactive Material Areas (RMAs) in an unescorted manner. If your work assignment requires access to areas posted for higher radiological hazards like Radiological Buffer Areas (RBAs), Radiation Areas, Contamination Areas, and others you are required to have Rad Worker training. Once you complete this training you will be qualified to access RCAs and RMAs at LANL without an escort. However, you are not qualified to operate radiation producing devices, work with radioactive material, or do work where you may receive greater than 100 mrem of exposure in a year.

61 RADIATION PROTECTION AND DOSIMETRY↗